1. Field of Invention
The field of invention is neutron detectors. More particularly, the present invention relates to a neutron detector having solid absorbers.
2. Description of the Related Art
Neutron detectors are useful in many applications. One such application is the monitoring of fissile material in storage containers, in spent nuclear fuel and in waste. Limiting the unauthorized transfer of fissile material is endorsed by all countries participating in the International Atomic Energy Agency (IAEA) program. The purpose of the IAEA program is to limit access to those materials needed for constructing a nuclear weapon of mass destruction.
A major effort is being made to provide robust neutron detection instrumentation systems at all locations where fissile material from spent nuclear fuel and excess plutonium from obsolete nuclear weapons are stored. Quantitative measurements are required for nuclear accountability of the fissile mass placed in criticality-safe storage containers. The 240Pu isotope is present in these materials and is a spontaneous emitter of about 1000 neutrons/second/gram of the isotope. Fissile material emits high energy gamma radiation and this, combined with the measurement of neutrons, allows real-time monitoring of the stored material.
Fissile material containing 235U and 239Pu can be detected by the measurement of excess neutrons when materials containing these isotopes are subjected to an external source of neutrons. The measurement techniques include the differential die-away technique, the Californium shuffler and the AmLi add-a-source. Typical of the art are those techniques described by Phillip Rinard, “Calculating Shuffler Count Rates,” LA-13815-MS, August 2001 and Greg Becker et al., “Transuranic and Low-Level Boxed Waste Form Nondestructive Assay Technology Overview and Assessment,” INEEL/EXT-99-00121, February 1999. Each technique is principally developed by the Los Alamos National Laboratory.
Another industry in which neutron detectors are widely used is the oil industry, in which neutron detectors are used to detect potential oil yielding sites. Oil producing formations deep in the earth emit neutrons at a different rate than water bearing formations or non-fluid bearing rock. A device using neutron detection for logging oil wells is disclosed in U.S. Pat. No. 4,641,028 issued to Taylor et al., on Feb. 3, 1987.
The '028 patent teaches a well logging instrument for use in a cased well bore. The '028 device contains a sealed source of fast neutrons and two identical thermal neutron detectors with a volume of four atmospheres of 3He gas. The formation surrounding the cased well bore is bombarded with high energy neutrons and the two thermal neutron detectors are spaced apart from one another and from the source to receive slowed down or thermal neutrons from the surrounding formations. The epithermal or fast neutrons striking the formation are slowed down by fluids containing great quantities of hydrogen or chlorine atoms, creating thermal or slow neutrons, which the detectors respond to logarithmically and independently. The counting rate of each detector is processed independently to count rate meters and to a recorder to present two outputs of information. The presence or absence of hydrogen and chlorine atoms in the formations is detected by each detector.
U.S. Pat. No. 5,532,482 issued to Stephenson on Jul. 2, 1996, teaches a method for determining a characteristic of an underground formation. The '482 method includes the steps of irradiating the formation with high-energy neutrons and detecting neutrons scattered by the formation. The detected neutrons have energies above epithermal to determine the nature of the formation matrix. Stephenson teaches that epithermal neutrons can also be detected to determine formation porosity. Stephenson utilizes an apparatus having a high-energy neutron source, typically a D-T accelerator producing 14 MeV neutrons, and detectors such as 4He-filled proportional counters for detecting neutrons having energies above epithermal and 3He-filled proportional counters for detecting epithermal neutrons.
In addition to the oil industry, neutron detectors are also commonly used in the medical industry. Neutron detectors are also useful for surveillance in nuclear facilities and weapons storage. While several specific utilities of neutron detectors are mentioned, it is well known to those skilled in the neutron detection art that neutron detectors are useful in many applications.
Neutrons are uncharged particles that can travel through matter without ionizing the matter. Because neutrons travel through matter in such a manner, they are difficult to detect directly. Some other evidence of a neutron event must be detected in order to determine its existence. An indirect method detects the result of a neutron event and not the neutron event itself.
The use of indirect detection of neutrons is known in the art. For example, a neutron detector as disclosed in U.S. Pat. No. 5,334,840 issued to Newacheck et al., on Aug. 2, 1994. The '840 neutron detector detects photons of light emitted by carbon infiltrated boron nitride in its hexagonal form when the compound is bombarded by neutrons. The amount of light detected correlates to the number of neutrons bombarding the boron nitride.
Another neutron detector commercially available utilizes 3He as the neutron absorber, such as in the '028 device described above. When bombarded by neutrons, 3He decomposes into H and H3 having combined kinetic energies of 764 keV. The ionization of the gas electrons can be detected using conventional methods well known in the art and further described below. This type of neutron detector requires a long collection time for the resulting ionization, requiring integrating and differentiating time constants of between 1 and 5 microseconds for the best results.
Other gas mixtures are commercially available that have varying resolution or charge per pulse yields depending on the gases used.
Neutron detection for monitoring the dose of thermal neutrons given to patients receiving boron neutron-capture therapy has used 6Li and a cerium activator in a glass fiber. See M. Bliss et al., “Real-Time Dosimetry for Boron Neutron-Capture Therapy,” IEEE Trans. Nucl. Sci., Vol. 42, No. 4, 639-43 (1995). Hiller et al., in U.S. Pat. No. 5,973,328, issued on Oct. 26, 1999, improve this technique by allowing a cerium-activated glass fiber to be coated with fissionable elements. A wet chemistry method of placing radioactive fissile elements into glass—which in the vitrified state does not pose a hazard—as described in the '328 patent using sol-gel based technology, is a significant benefit. M. Ghioni et al., “Compact Active Quenching Circuit for Fast Photon Counting with Avalanche Photodiodes,” Rev. Sci. Instr., 67, 3440-48 (1996), describe an avalanche photodiode implementation for detecting neutron induced ionization and optical pulse detection.
The '328 device introduced sol-gel techniques unique in the art of neutron detection. Sol-gel chemistry was first discovered in the late 1800s. This area of chemistry received renewed interest when the process was found useful in producing monolithic inorganic gels at low temperatures that could be converted to glasses without a high temperature melting process. C. J. Brinker et al., “Sol-Gel Science: The Physics and Chemistry of Sol-Gel Processing” (Academic Press, Inc., New York 1990) provide a comprehensive explanation of sol-gel chemistry. Sheng Dai et al., “Spectroscopic Investigation of the Photochemistry of Uranyl-Doped Sol-Gel Glasses Immersed in Ethanol,” Inorg. Chem., 35, 7786-90 (1996), provide further detail disclosing uranyl-doped sol-gel glasses.
Emissions detectors such as microchannel plates, channeltrons, and avalanche photodiodes (APDs) are in common use for detecting ultraviolet (UV) light and fissioned charged particles such as electrons or protons. Microchannel plates are commercially available and well known in the art. Typically a microchannel plate is formed from lead glass having a uniform porous structure of millions of tiny holes or microchannels. Each microchannel functions as a channel electron multiplier, relatively independent of adjacent channels. A thin metal electrode is vacuum-deposited on both the input and output surfaces to electrically connect channels in parallel. Microchannel plates can be assembled in stacked series to enhance gain and performance.
The microchannel plates serve to amplify emissions from fissionable material resulting from the bombardment of neutrons. The amplified signal is then detected and recorded. The signal frequency is proportional to the charged particle emissions, which are proportional to the amount of neutrons bombarding the fissionable material.
Typically due to the exotic materials and sensitivity of the equipment, the neutron detectors currently available are expensive and difficult to maintain. For example 3He is an extremely rare stable isotope and must be separated at considerable expense from the radioactive gas tritium. Furthermore, the use of a gas absorber results in a slower response time than a solid absorber as disclosed herein. The '328 device thus incorporates fissionable material into a sol-gel composition in combination with an emission detector.
Neutron scattering is a powerful tool for conducting scientific studies of the physical geometry of molecules important in biology and material science. Protein structure and the structure of superconductors are of immediate practical importance. The detection of the scattered neutrons is an area where advancements are necessary. Specifically of concern are the ability to locate the scattered neutrons and the ability to rapidly process the neutron signals as the detection rate becomes very rapid.
A major research facility using spallation constructed by the Department of Energy is the Spallation Neutron Source (SNS) facility in Oak Ridge, Tenn. Spallation is a nuclear reaction in which incident particles bombard an atomic nucleus to eject particles from the nucleus. The SNS is designed to have an output of pulsed neutrons that is the most intense in the world of its type. The SNS is provided with multiple experimental stations using pulsed neutrons. One such experimental station locates the neutrons diffracted from a target on an x-y plane. The time of the neutron absorption on the x-y plane surface is then referenced to a fiducial timing signal to an accuracy of within 100 nanoseconds of the absorption event.
Mori et al., “Measurement of Neutron and γ-ray Intensity Distributions with an Optical Fiber-Scintillator Detector,” Nuclear Instruments and Methods in Physics Research, A 422, 129-132 (1999), describe a ZnS(Ag) scintillator with 6Li on the tip of an optical fiber for locating neutrons with a position resolution of 1 mm in a 10 minute interrogation and within a volume wherein the tip is extendable one meter. Gorin et al., “Development of Scintillation Imaging Device for Cold Neutrons,” Nuclear Instruments and Methods in Physics Research, A 479, 456-460 (2002), have described using ZnS(Ag) and 6LiF coupled to wavelength-shifting fibers. Gorin et al., employ arrays of fibers in two planes rotated ninety degrees relative to each other such that an absorbed neutron is located in the plane to a resolution of 1 mm.
Wallace et al., Nuclear Instruments and Methods A 483 (2002) 764-773 report the gamma insensitivity of the thin film lithiated glass. This specificity of the lithiated glass for generating a signal in the presence of gamma radiation has application to the monitoring of spent nuclear fuel rods and for the determination of the fissile mass within remotely handled transuranic waste.
Other methods and devices have been developed for neutron detection. Typical of the art are those methods and devices disclosed in the following U.S. patents and patent applications:
Patent No.Inventor(s)Issue Date3,222,521K. EinfeldDec. 7, 19654,365,159C. A. YoungDec. 21, 19824,481,421C. A. Young et al.Nov. 6, 19845,289,510J. T. MihalczoFeb. 22, 19945,336,889K. J. HofstetterAug. 9, 19945,345,084R. C. ByrdSept. 6, 19945,659,177R. L. Schulte et al.Aug. 19, 19975,726,453R. G. Lott et al.Mar. 10, 19985,880,471J. Schelton et al.Mar. 9, 19995,968,425A. Bross et al.Oct. 19, 19995,973,328J. M. Hiller et al.Oct. 26, 19996,134,289A. J. Peurrung et al.Oct. 17, 20006,218,670J. C. Yun et al.Apr. 17, 2001
Of these patents, Einfeld ('521) teaches a method and apparatus for the non-destructive testing of a substance to determine the concentration of two or more fissionable isotopes in the substance. Einfeld teaches generation of first and second neutron spectrums, each having a unique mean energy. Determination of the number of fissions as a function of the neutron spectrum applied follows from the counting of the prompt and/or delayed neutrons produced by the fissions.
Young ('159) teaches a neutron detection apparatus including a selected number of flat surfaces of 6Li foil. A gas mixture is in contact with each of the flat surfaces for selectively reacting to charged particles emitted by or radiated from the lithium foil. A container is provided to seal the lithium foil and the gas mixture in a volume from which water vapor and atmospheric gases are excluded, the container having one or more walls capable of transmitting neutrons. Monitoring equipment in contact with the gas mixture detects reactions taking place in the gas mixture and, in response to such reactions, provides notice of the flux of neutrons passing through the volume of the detector.
Similarly, Young et al., ('421) teach a neutron detection apparatus is provided including a selected number of surfaces of 6Li coated wire mesh in contact with a gas mixture for selectively reacting to charged particles emitted or radiated by the 6Li coated mesh. As in the '159 device, a container is provided to seal the 6Li coated mesh and the gas mixture in a volume from which water vapor and atmospheric gases are excluded, the container having one or more walls capable of transmitting to neutrons. Monitoring equipment in contact with the gas mixture detects the generation of charged particles in the gas mixture and, in response to such charged particles, provides an indication of the flux of neutrons passing through the volume of the detector.
In the '510 patent, Mihalzco teaches nuclear reaction detectors capable of position sensitivity with sub-millimeter resolution in two dimensions. The nuclear reaction detectors include two arrays of scintillation or wavelength shifting optical fibers. Each array is formed of a plurality of optical fibers disposed in a side-by-side relationship. The two arrays are disposed in X- and Y-directions with respect to each other, with a layer of nuclear reactive material disposed between and operatively associated with surface regions of the optical fiber arrays. Each nuclear reaction occurring in the layer of nuclear reactive material produces energetic particles for simultaneously providing a light pulse in a single optical fiber in the X-oriented array and in a single optical fiber in the Y-oriented array. These pulses of light are transmitted to a signal producing circuit for providing signals indicative of the X-Y coordinates of each nuclear event.
The nuclear reactive material of the '510 patent is doped with a phosphor such as calcium tungstate, magnesium tungstate, zinc silicate, zinc sulfide, cadmium tungstate, and cadmium borate. Mihalzco further teaches that a compound such as 6LiF or glass or plastic scintillators containing 235U, 10B, or 238U provides a concentration of the phosphor dopant in the layer of nuclear reactive material sufficient to assure that an adequate distribution of phosphor to be contacted by and react with the energetic particle is produced from each nuclear reaction. Normally, a concentration of the phosphor dopant in the range of about 100 ppm to about 2 percent by volume is adequate for the purposes of the Mihalzco detectors.
Hofstetter ('889) discloses a gamma radiation detector using a radioluminescent composition. The detector includes a radioluminescent composition that emits light in a characteristic wavelength region when exposed to γ radiation. The composition contains a scintillant such as anglesite (PbSO4) or cerussite (PbCO3) incorporated into an inert, porous glass matrix via a sol-gel process. Particles of radiation-sensitive scintillant are added to a sol solution. The mixture is polymerized to form a gel, and then dried under conditions that preserve the structural integrity and radiation sensitivity of the scintillant. The final product is a composition containing the uniformly-dispersed scintillant in an inert, optically transparent and highly porous matrix. Hofstetter describes the resulting composition as chemically inert and substantially impervious to environmental conditions.
In the Byrd ('084) device, a plurality of omnidirectional radiation detectors is arranged in a closely packed symmetrical pattern to form a segmented detector. The output radiation counts from these detectors are arithmetically combined to provide the direction of a source of incident radiation. Output counts from paired detectors are subtracted to yield a vector direction toward the radiation source. The counts from all of the detectors are combined to yield an output signal functionally related to the radiation source strength.
R. L. Schulte et al., ('177) teach a directional thermal neutron detector for detecting thermal neutrons and determining the direction of the thermal neutron source. The directional detector includes an array of individual thermal neutron detector modules, each of which comprises front and back planar silicon detectors between which is disposed a gadolinium foil. The array comprises a plurality of individual detector modules angularly displaced with respect to each other. The direction of the thermal neutron source is determined by comparing the magnitudes of the output signals from the plurality of angularly displaced detector modules. Each thermal neutron detector module is segmented into four quadrants to reduce its capacitance and resultant noise. The thickness of the gadolinium foil in each thermal neutron detector module is at least 15 microns thick, to improve the front-to-back silicon detector counting ratio to ascertain the side (front or back) from which thermal neutrons are arriving at the detector. The thick gadolinium foil makes each detector module substantially opaque to thermal neutrons, and the detector modules are positioned relative to each other in the array to shield one another from thermal neutrons, thereby enhancing the angular resolution of the directional thermal neutron detector. Gamma rays are discriminated against by using coincidence signal processing within the elements of the detector sandwich to reduce the gamma ray contribution to the total signal.
Lott et al., ('453) disclose a radiation resistant solid state neutron detector. The '453 detector uses a neutron converter material such as boron or lithium to react with neutrons to create charged particles that are received in a semiconductor active region of the detector. The active thickness of the detector is smaller than the range of the charged particles. Since most of the radiation damage produced by impinging charged particles occurs near the end of the range of the particles, displacement damage predominantly occurs outside of the active region. Although the charged particles pass through the semiconductor material, the particles cause electron excitation within the semiconductor material, the electron excitation being detected in the form of an electronic pulse. The '453 detector is provided to increase resistance to radiation damage, improve high temperature operation, and to obtain real time measurements of neutron flux in reactor cavities and other previously inaccessible locations.
Schelton et al., ('471) disclose a neutron detector for the detection of thermal neutrons. The '471 neutron detector includes 6LiF layers for the conversion of the neutrons to ionizing radiation. The 6LiF layers are surrounded by layers for detecting the ionizing radiation generated by the neutrons in the 6LiF layers.
The '425 patent issued to Bross et al., discloses methods for the continuous production of the plastic scintillator material. The methods employ either two major steps (tumble-mix) or a single major step (inline-coloring or inline-doping). Using the two step method, the polymer pellets are mixed with silicone oil, and the mixture is then tumble mixed with the dopants necessary to yield the proper response from the scintillator material. The mixture is then placed in a compounder and compounded in an inert gas atmosphere. The resultant scintillator material is then extruded and pelletized or formed. When only a single step is employed, the polymer pellets and dopants are metered into an inline-coloring extruding system. The mixture is then processed under a inert gas atmosphere, usually argon or nitrogen, to form plastic scintillator material in the form of either scintillator pellets, for subsequent processing, or as material in the direct formation of the final scintillator shape or form
Peurrung et al., ('289) teach a system for measuring a thermal neutron emission from a neutron source. The '289 device includes a reflector/moderator proximate the neutron source that reflects and moderates neutrons from the neutron source. The reflector/moderator further directs thermal neutrons toward an unmoderated thermal neutron detector.
U.S. Pat. No. 5,973,328 issued to J. M. Hiller et al., discloses a neutron detector composed of fissionable material having ions of lithium, uranium, thorium, plutonium, or neptunium, contained within a glass film fabricated using a sol-gel method combined with a particle detector. When the glass film is bombarded with neutrons, the fissionable material emits fission particles and electrons. The '328 patent further discloses prompt emitting activated elements yielding a high energy electron contained within a sol-gel glass film in combination with a particle detector. The emissions resulting from neutron bombardment can then be detected using standard UV and particle detection methods well known in the art, such as microchannel plates, channeltrons, and silicon avalanche photodiodes.
Currently pending U.S. Patent Application 2003/0178574 filed by the inventor of the present application, along with A. Stephan, S. Dai and H. J. Im, discloses a neutron detector composed of a matrix of scintillating particles imbedded in a lithiated glass. The neutron detector detects the neutrons by absorbing the neutron in the 6Li isotope which has been enriched from the natural isotopic ratio to a commercial ninety five percent. The utility of the '574 detector is optimized by suitably selecting scintillating particle sizes in the range of the alpha and the triton. Nominal particle sizes are in the range of five to twenty five microns depending upon the specific scintillating particle selected.
Other references of interest in the art of neutron detection include:    H. Krinninger et al., “Pulsed Neutron Method for Non-Destructive and Simultaneous Determination of the 235U and 239Pu Contents of Irradiated and Non-Irradiated Reactor Fuel Elements,” Nucl. Instr. Meth. 73, 13-33 (1969);    M. Zanarini et al., “Evaluation of Hydrogen Content in Metallic Samples by Neutron Computed Tomography,” IEEE Trans. Nucl. Sci., 42, 580-84 (1995);    C. M. Logan et al., “Observed Penetration of 14-MeV Neutrons in Various Materials,” Nucl. Sci. Eng. 115, 38-42 (1993);    H. Jaeger et al., “Two-Detector Coincidence Routing Circuit for Personal Computer-Based Multichannel Analyzer,” Rev. Sci. Instrum. 66, 3069-70 (1995);    E. J. T. Burns et al., “A Solenoidal and Monocusp Ion Source (SAMIS),” Rev. Sci. Instr., 67, 1657-60 (1996);    S. T. Coyle et al., “A Low Cost Preamplifier for Fast Pulses From Microchannel Plates,” Rev. Sci. Instr., 66 4000-01 (1995);    Y. G. Kudenko et al., “Extruded Plastic Counters with WLS Fiber Readout,” Nucl. Inst. And Meth. A 469, 340-346 (2001);    C1207-97 Standard Test Method for Nondestructive Assay of Plutonium in Scrap and Waste by Passive Neutron Coincidence Counting, ASTM International;    W. Harker et al., “Demonstration Neutron Multiplicity Counter Coincidence Counting Software for Authentication,” Los Alamos National Laboratory Report LA-UR-01-4186, July 2001;    R. Hogle et al., “APNEA list mode data acquisition and real-time event processing,” 5th Nondestructive Assay and Nondestructive Examination Waste Characterization Conference, January 1-16, Salt Lake City, Utah, 1997;    B. D. Lebedev et al., “Monte Carlo Calculation to Optimize the Neutron Multiplicity Counter for Measurement of Representative Plutonium Items in AT 400 Container,” Proceedings of the INMM 43rd Annual Meeting, 2002;    S. Croft et al., “Principles of Fast Neutron Detector Package Design for Differential Dieaway Technique Assay,” Proceeding of the INMM 43rd Annual Meeting, 2002;    Yun Chan Kang et al., “Y2SiO5:Ce Phosphor Particles 0.5-1.4 micrometer in Size with Spherical Morphology,” J. Solid State Chem., 146 (1999) 168-175;    A. P. Bartkoetal et al., “Observation of dipolar emission patterns from isolated Eu3+:Y2O3 doped nanocrystals: new evidence for single ion luminescence,” Chemical Physics Letters 358 (2000) 459-465; and    J. Y. Choe et al., “Cathodluminescence study of novel sol-gel derived Y3-x Al5O12:Tbx phosphors,” Journal of Luminescence 93 (2001) 119-128.
The trafficking in fissile material capable of being fabricated into a nuclear weapon of mass destruction is publicized to be so profitable that significant resources are being expended to prevent the diversion of material for clandestine sales. Iraq has been the focus of international attention because that government has actively sought to acquire fissile material. Other nations in the Mid East may want to clandestinely acquire fissile material so as to have the option of developing a nuclear weapon in the future. Such a clandestine hording of material would be virtually impossible to detect if placed in cold storage without the generation of new structures that would be detected by satellite surveillance. Neutron detectors offer the means to observe fissile material using active and passive measurement techniques.
Two attributes are necessarily measured in all systems used for the unambiguous identification of fissile material. These include gamma radiation and the active and passive techniques of measuring for the presence of neutrons.
The development of systems for measuring gamma rays starting at 59.5 keV associated with Americium and extending up to 414 keV for Plutonium is mature. The Los Alamos National Laboratory and the Lawrence Livermore National Laboratory have developed sodium iodide (NaI) and high purity germanium (HPGE) gamma ray detection systems that are field portable and operator friendly for measuring radiation in the desert. Large neutron detectors have not had as much effort expended in the development of field detectors. Improved technology exists to provide a solid-state neutron detector for field measurement of neutrons using aerial surveys. The same detector can be truck mounted allowing surveys having a standoff from the buildings being observed.
The most useful rapid means of surveying a large area for radiation is an aerial survey using a helicopter. A large volume neutron detector that can be towed below a helicopter allows a rapid survey to be conducted of any facility and the surrounding buildings without a need to place personnel with instruments at risk. However, large neutron detectors have not been developed for this application. The same detectors could be placed into trucks and any activity seen in the air can be confirmed by having surveys made around the perimeter of the hot buildings.
With reference again to the '328 patent, which discloses the means for producing a glass containing a high loading of 6Li, the glass is pipetted upon the surface of a silicon charged particle detector and the triton and alpha particles from neutron absorption enter the detector signaling the presence of neutrons. The enhancement on the technology takes the original technique of generating a lithiated glass using a sol-gel process and couples it with the manufacture of a composite material where micron size organic and inorganic scintillating particles are embedded within the glass. This method has been demonstrated. A further development of the technology is the placement of a thin 1 mm layer of this scintillating composite between two arrays of orthogonal wavelength shifting fibers. The basic 2-D imaging technique using a scintillation layer between wavelength shifting fibers is reported by Gorin, A., et al., Nuclear Instruments and Methods A 479, 456 (2002).
Using high speed coincidence circuitry based upon positron emission tomography the location of neutron absorption can be located on two of the fibers allowing a 1 mm resolution on a tile of one square meter. The scintillation output in the inorganic particles is roughly 100,000 photons. The spherical wave from the ionization trail in the particle couples to a fiber above and a fiber below the glass film and the x-y coordinate is recorded in a high speed data acquisition system. Additionally the time of the detection of a neutron is to be made to within 100 nanoseconds of a fiducial timing signal. Multiple detectors are to be tiles so that at about thirty square meters of detector area are monitored. The final product is a system to be used at the Spallation Neutron Source for making neutron diffraction measurements of cold neutrons.
Specific reference is made to the MINOS neutrino oscillation detector under construction at Fermi National Accelerator Laboratory. In particular, reference is made to http://www-numi.fnal.gov/minwork/info/minos_tdr.html, Chapter 5, Scintillator detector fabrication detailing the MINOS scintillator system. Sections 5.1 through 5.5 which is a detector which uses an extruded plastic scintillator read out by wavelength shifting (WLS) fibers coupled to multi-pixel photodetectors. The polystyrene used for the MINOS detector contains a fluor at a 175 ppm doping concentration.
Coincidence detection of neutrons can be automated into hardware as can be licensed from Los Alamos National Laboratory. In an alternate method, every neutron that is in every neutron detector in a plurality of detectors with respect to fiducial timing markers is recorded. Post-processing the millions of detected neutrons that are taken in making a fissile material mass measurement has been developed for field measurements of transuranic waste destined for the Waste Isolation Pilot Plant in Carlsbad, N. Mex. The entire list of events is sorted in a few minutes for tagging the spontaneous neutron events and by use of calibration standards giving the fissile waste mass in a fifty-five gallon waste drum. The same coincidence method is used for small one and five gallon containers used for storing enriched uranium oxide and plutonium oxide. These systems are based upon using 3He tubes containing the gas at a pressure of three or four atmospheres embedded in polyethylene walls surrounding the material being measured for fissile mass. The polyethylene is thermalizing the neutrons which then enter the 3He tubes. The tubes may be concentric layers up to three deep to achieve high capture efficiency. The present invention proposes to replace the polyethylene with polystyrene and the 6Li glass/scintillation particulates for the 3He tubes. A key figure that is a measure of the power of the system to design the dimensions of walls and the placement of the 3He tubes in such a way as to thermalize the neutrons and capture them in the shortest interval after their creation. The use of polystyrene as the matrix for placement of the 6Li glass/scintillating particulates allows great design flexibility in the selection of the density per cubic centimeter to minimize the time for thermalization and neutron capture for differing enclosed volumes.